A Benchmark Experiment for Fast Neutron Transport in Graphite

Author(s):  
H. Takahashi ◽  
K. Sugiyama
Author(s):  
Cécile-Aline Gosmain ◽  
Sylvain Rollet ◽  
Damien Schmitt

In the framework of surveillance program dosimetry, the main parameter in the determination of the fracture toughness and the integrity of the reactor pressure vessel (RPV) is the fast neutron fluence on pressure vessel. Its calculated value is extrapolated using neutron transport codes from measured reaction rate value on dosimeters located on the core barrel. EDF R&D has developed a new 3D tool called EFLUVE3D based on the adjoint flux theory. This tool is able to reproduce on a given configuration the neutron flux, fast neutron fluence and reaction rate or dpa results of an exact Monte Carlo calculation with nearly the same accuracy. These EFLUVE3D calculations does the Source*Importance product which allows the calculation of the flux, the neutronic fluence (flux over 1MeV integrated on time) received at any point of the interface between the skin and the pressure vessel but also at the capsules of the pressurized water reactor vessels surveillance program and the dpa and reaction rates at different axial positions and different azimuthal positions of the vessel as well as at the surveillance capsules. Moreover, these calculations can be carried out monthly for each of the 58 reactors of the French current fleet in challenging time (less than 10mn for the total fluence and reaction rates calculations considering 14 different neutron sources of a classical power plant unit compared to more than 2 days for a classic Monte Carlo flux calculation at a given neutron source). The code needs as input: - for each reaction rate, the geometric importance matrix produced for a 3D pin by pin mesh on the basis of Green’s functions calculated by the Monte Carlo code TRIPOLI; - the neutron sources calculated on assemblies data (enrichment, position, fission fraction as a function of evolution), pin by pin power and irradiation. These last terms are based on local in-core activities measurements extrapolated to the whole core by use of the EDF core calculation scheme and a pin by pin power reconstruction methodology. This paper presents the fundamental principles of the code and its validation comparing its results to the direct Monte Carlo TRIPOLI results. Theses comparisons show a discrepancy of less than 0,5% between the two codes equivalent to the order of magnitude of the stochastic convergence of Monte Carlo results.


1972 ◽  
Author(s):  
R. E. Maerker ◽  
F. J. Muckenthaler ◽  
C. E. Clifford

1985 ◽  
pp. 849-855
Author(s):  
J. Burian ◽  
B. Janský ◽  
M. Marek ◽  
J. Rataj

Author(s):  
Choon Sung Yoo ◽  
Byoung Chul Kim ◽  
Tae Je Kwon

A Pressurized Thermal Shock (PTS) Event is an event or transient in pressurized water reactors (PWRs) causing severe overcooling (thermal shock) concurrent with or followed by significant pressure in the reactor vessel. A PTS concern arises if one of these transients acts in beltline region of a reactor vessel where a reduced fracture resistance exists due to neutron irradiation. Such an event may produce a flaw or cause the propagation of a flaw postulated to exist near the inner vessel wall surface, thereby potentially affecting the integrity of the vessel. In this paper fast neutron flux reduction techniques were implemented to reduce the potential risk of PTS due to the neutron irradiation on the pressure vessel beltline region. And the RTPTS value for the end of life of the plant was projected using the fast neutron fluence obtained by neutron transport calculations according to the various core loading pattern and reduction program possible for the future cycles.


1995 ◽  
Vol 28 ◽  
pp. 745-752 ◽  
Author(s):  
Chikara Konno ◽  
Fujio Maekawa ◽  
Yukio Oyama ◽  
Yujiro Ikeda ◽  
Hiroshi Maekawa

1972 ◽  
Vol 47 (3) ◽  
pp. 329-348 ◽  
Author(s):  
B. K. Malaviya ◽  
N. N. Kaushal ◽  
M. Becker ◽  
E. T. Burns ◽  
A. Ginsberg ◽  
...  

2018 ◽  
Vol 169 ◽  
pp. 00009
Author(s):  
Toni Kögler ◽  
Roland Beyer ◽  
Arnd R. Junghans ◽  
Ronald Schwengner ◽  
Andreas Wagner

The fast-neutron-induced fission cross section of 242Pu was determined in the energy range of 0.5 MeV to 10MeV at the neutron time-of-flight facility nELBE. Using a parallel-plate fission ionization chamber this quantity was measured relative to 235U(n,f). The number of target nuclei was thereby calculated by means of measuring the spontaneous fission rate of 242Pu. An MCNP 6 neutron transport simulation was used to correct the relative cross section for neutron scattering. The determined results are in good agreement with current experimental and evaluated data sets.


After two major nuclear power plant accidents in Chernobyl (1986) and Fukushima (2011), one of the main requirements for the nuclear power engineering is the safety of the nuclear reactors in operation, as well as new nuclear power plants of the fourth generation, which are being developed now. One of such requirements is presence of the so-called “inherent safety” mechanism, which renders the uncontrolled reactor runaway impossible under any conditions, moreover, the implementation of such a mechanism should be ensured on the level of physical principles embedded in the reactor design. Another important problem of the nuclear power engineering is the need of the transition to the large-scale use of the fast-neutron breeder reactors, with which it would be possible to set up expanded reproduction of the nuclear fuel and by that means solve the problem of supplying humanity with relatively cheap energy for thousands of years. Moreover, at present an unresolved problem is the disposal of spent nuclear fuel containing radioactive nuclides with long half-lives, which presents a long-term danger to the ecology. One of the promising conceptions of the fast-neutron breeder reactor, which can, in the case of successful implementation, partially or even entirely solve the problems of the nuclear power engineering mentioned above, is the reactor that operates in the nuclear burning wave mode, which is also known as “Traveling wave reactor”, CANDLE and by some other names. This paper presents a short review of the main theoretical approaches used for description of such a physical phenomenon as slow nuclear burning (deflagration) wave in the neutron multiplication medium initially composed of the fertile material 238U or 232Th. A comparative analysis of the possibilities of different mathematical models for describing this phenomenon is performed, both for those based on the deterministic approach (i.e. solving neutron transport equations) and for models that use Monte Carlo methods. The main merits of the fast breeder reactor, working in the nuclear burning wave mode, and problems related to the practical realization of the considered concept are discussed.


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